1. Field of the Invention
The present invention relates to a zirconium alloy material.
2. Description of Related Art
Zirconium alloys are used as materials for chemicals manufacturing apparatuses/devices due to their satisfactory corrosion resistance. In addition, zirconium alloys are used as materials for core internals of nuclear reactors such as fuel cladding tubes and channel boxes due to their small thermal neutron cross-sections.
For example, zirconium alloys each containing 1.2 to 1.7 percent by mass of tin (Sn) such as R60802 and R60804 alloys prescribed in American Society for Testing and Materials Standards (ASTM) B811, and ZrTN802D and ZrTN804D alloys prescribed in Japanese Industrial Standards (JIS) H4751 are used in nuclear fuel cladding tubes, spacers, and channel boxes of light-water reactors typified by pressurized water reactors and boiling water reactors. Independently, zirconium alloys each containing 2.0 to 3.0 percent by mass of niobium (Nb) such as R60901 alloy prescribed in ASTM B811 are used in pressure tubes typically of heavy-water reactors.
In use environments of materials for core internals, even zirconium alloy materials undergo corrosion which proceeds during a long-term use. This causes the zirconium alloy materials to have a thick film to thereby have a lower heat transfer coefficient.
In a fuel rod, the fracture of the cladding tube may occur because temperature rise of uranium pellets accelerates the emission of fission product gases (fission products (FPs) in a gaseous form) to increase the inner pressure of the cladding tube. In addition, progression of the corrosion may increase the amount of hydrogen absorption, and this may cause hydrogen embrittlement. For these reasons, the corrosion is considered to be one of factors which decide lives of instruments composed of the zirconium alloys.
Zirconium alloys for constituting such instruments are demanded to have further higher corrosion resistance so as to use the instruments for a long time. To meet this demand, improvements such as those in manufacturing process and in chemical composition have been investigated in order to improve the corrosion resistance of zirconium alloys.
In a zirconium alloy, there are an α-phase (hexagonal closest packing structure: HCP) region stable at room temperature; a β-phase region stable at high temperatures of about 950° C. or higher, of which transformation temperature may vary depending on the chemical composition; and an α+β phases region between these regions.
In a manufacturing process of a zirconium alloy material, a billet prepared through melting and subsequent forging is subjected to a solution heat treatment called β-quenching, in which the material is held in a single p-phase region and then quenched, in order to improve corrosion resistance and to homogenize the whole material. After the β-quenching, the zirconium alloy material is subjected to a hot working according to necessity, followed by repeating cold workings with the interposition of annealing operations in the α+β phases region or α-phase region to form the material into a predetermined shape of the target member. In the solution heat treatment, the zirconium alloy material may be held in the α+β phases region before quenching.
Sn-containing zirconium alloys undergo precipitation of intermetallic compounds such as Zr(Cr,Fe)2 and Zr2(Ni,Fe). Nb-containing zirconium alloys undergo precipitation of β-zirconium or β-niobium. Nb-containing zirconium alloys further containing Sn and/or Fe undergo precipitation of intermetallic compounds such as (Zr,Nb)3Fe, (Zr,Nb)2Fe, Zr(Fe,Nb)2 and Zr(Fe,Nb)3. Formation and growth of these precipitates occurs during holding at high temperatures after β-quenching.
In thermoforming after β-quenching, working ratio and annealing conditions are controlled to be suitable conditions in consideration of the texture formation of alloy and the formation and growth of precipitates. The texture interacts typically with the irradiation growth and hydrogen embrittlement of the alloy, whereas the precipitates affect the corrosion resistance of the alloy. If the precipitates are coarsely grown, the corrosion resistance is adversely affected. To avoid this, the thermal history in thermoforming after β-quenching is controlled so as to avoid long-term holding of the alloy material at high temperatures.
B-quenching helps the alloy material to have better corrosion resistance if it is performed at a high cooling rate.
Japanese Unexamined Patent Application Publication No. 2001-262259 (Document 1) discloses a Sn-containing zirconium alloy in which liquid sodium is used instead of water as a cooling medium in β-quenching.
Japanese Unexamined Patent Application Publications No. S63-33535, No. S64-39589, No. H02-271291 and No. H10-273746 (Documents 2 to 5) disclose zirconium alloys each having a lower Sn content to suppress uniform corrosion to thereby improve corrosion resistance.
Japanese Patent No. 2600057 and Japanese Patent No. 2790138 (Documents 6 and 7) disclose Sn-containing zirconium alloys each having a higher Fe content and having a specific ratio of Fe content to Ni content in order to help the Sn-containing zirconium alloys to have better corrosion resistance.
Japanese Unexamined Patent Application Publications No. H11-101887 and No. H11-109072 (Documents 8 and 9) disclose Sn—Nb-containing zirconium alloys each including finely dispersed β-niobium grains having an average grain size of 30 to 150 nm. The zirconium alloys are obtained by performing a heat treatment at 540° C. for 16 hours after β-quenching, in order to provide high corrosion resistance even in an irradiation environment.
In a nuclear fuel cladding tube, an alloy material is generally subjected to final annealing in thermoforming, thereafter straightened using a roll straightener, and the outer surface of which is finished by mechanical polishing or acid pickling. Japanese Unexamined Patent Application Publications No. H11-52087 and No. 2000-105289 (Document 10 and 11) disclose zirconium alloy nuclear fuel cladding tubes which have been manufactured by performing shot peening on the surface and subsequently performing a heat treatment at 560° C. to 620° C. in final stages of their manufacturing, to form, in the outer surface, an ultrafine grain layer having an average grain size of 3 μm or less and a thickness of from 1 to 100 μm so as to improve corrosion resistance.